1. Field of the Invention
The present invention relates generally to fuel assemblies for a nuclear reactor and, more particularly, is concerned with a boiling water reactor (BWR) fuel assembly incorporating features which eliminate large cross-flows between fuel rod mini-bundles at upper spacer locations to avoid degradation of critical heat flux (CHF) performance.
2. Description of the Prior Art
Typically, large amounts of energy are released through nuclear fission in a nuclear reactor with the energy being dissipated as heat in the elongated fuel elements or rods of the reactor. The heat is commonly removed by passing a coolant in heat exchange relation to the fuel rods so that the heat can be extracted from the coolant to perform useful work.
In nuclear reactors generally, a plurality of the fuel rods are grouped together to form a fuel assembly. A number of such fuel assemblies are typically arranged in a matrix to form a nuclear reactor core capable of a self-sustained, nuclear fission reaction. The core is submersed in a flowing liquid, such as light water, that serves as the coolant for removing heat from the fuel rods and as a neutron moderator. Specifically, in a BWR the fuel assemblies are typically grouped in clusters of four with one control rod associated with each four assemblies. The control rod is insertable within the fuel assemblies for controlling the reactivity of the core. Each such cluster of four fuel assemblies surrounding a control rod is commonly referred to as a fuel cell of the reactor core.
A typical BWR fuel assembly in the cluster is ordinarily formed by a N by N array of the elongated fuel rods. The bundle of fuel rods are supported in laterally spaced-apart relation and encircled by an outer tubular channel having a generally rectangular cross-section. Examples of such fuel assemblies are illustrated and described in U.S. Pat. Nos. 3,689,358 to Smith et al and 3,802,995 to Fritz et al and Canadian Pat. No. 1,150,423 to Anderson et al, as well as in the patent applications cross-referenced above.
In a fuel assembly of this type the fuel rods in the central region of the bundle thereof may be undermoderated and overenriched. In order to remedy this condition by increasing the flow of moderator water through this region of the assembly, several arrangements have been proposed. In the Fritz et al patent, one or more elongated empty rods are substituted for fuel rods in the central region of the assembly. In the Anderson et al patent, an elongated centrally-disposed stiffening device with vertical water passageways is used in the assembly. In the above cross-referenced Barry et al, Doshi and Lui patent applications, an elongated centrally-disposed water cross is used in the assembly.
As disclosed in the aforementioned latter four cross-referenced applications, the water cross has a plurality of four radial panels, forming a cruciform water flow channel, which divide the fuel assembly into four, separate elongated compartments, with the bundle of fuel rods being divided into mini-bundles supported by axially displaced grids or spacers and upper and lower tie plates disposed in the respective compartments. The cruciform water flow channel provides a centrally-disposed cross-shaped path for the flow of subcooled neutron moderator water within the channel along the lengths of, but separate from, adjacent fuel rods in the mini-bundles thereof.
In the above cross-referenced applications, the radial panels of the water cross are interconnected to the sides of the outer flow channel to support the water cross. However, it is considered advantageous to provide some means to permit coolant flow transversely between the separate minibundles of fuel rods of the fuel assembly to provide hydraulic pressure equalization therebetween. For example, in the Taleyarkhan application, pressure equalization openings permitting cross-flow between the compartments are defined between vertically spaced ribs formed in the sides of the outer flow channel which connect to the outer edges of the water cross panels.
Notwithstanding the improvements fostered by the cruciform water cross flow channel of the above cross-referenced applications with respect to hydraulic pressure equalization between the fuel rod mini-bundles in the separate compartments, other problems have recently been recognized which, if left unresolved, will cause degradation in CHF performance. These problems relate to certain of the openings which allow cross-flow communication of two phase (steam/water) coolant between the mini-bundles in order to achieve the desired flow stability and pressure equalization. At the locations of the fuel rod mini-bundle spacers, this cross-flow between the compartments can rise to large values and degrade CHF performance. The CHF phenomenon in BWRs is of paramount importance in characterizing the power rating of the plant. Hence degradation of this margin should be minimized.
Consequently, the need exists for further improvement of the BWR fuel assembly to eliminate or minimize CHF margin penalties and uncertainties.